NLS2014029, Relief Request from Certain Inservice Inspection Code Requirements Pursuant to 10 CFR 50.55a
| ML14202A081 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 07/15/2014 |
| From: | Limpias O Nebraska Public Power District (NPPD) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NLS2014029, TAC ME3319 | |
| Download: ML14202A081 (7) | |
Text
N Nebraska Public Power District Always there when you need us 10 CFR 50.55a NLS2014029 July 15, 2014 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001
Subject:
10 CFR 50.55a Request Number RI-08, Revision 0 Cooper Nuclear Station, Docket No. 50-298, DPR-46
Reference:
Cooper Nuclear Station - Request for Relief No. RI-04 for the Fourth 10-Year Inservice Inspection Interval Regarding Inspection of Reactor Vessel Nozzle-to-Vessel Shell Welds (TAC No. ME3319), dated October 8, 2010
Dear Sir or Madam:
The purpose of this letter is to request that the Nuclear Regulatory Commission (NRC) grant Nebraska Public Power District (NPPD) relief from certain inservice inspection (ISI) code requirements for Cooper Nuclear Station (CNS) pursuant to 10 CFR 50.55a.
10 CFR 50.55a Request Number RI-08, Revision 0 is applicable to the fourth ten-year ISI interval, which began March 1, 2006. NPPD requests NRC approval of the attached request by July 15, 2015, which represents a twelve month review period following the submittal.
RI-08, Revision 0 is contained in the attachment to this letter. The request is similar to the submittal the NRC Staff has approved in a Safety Evaluation referenced above for other nozzle-to-vessel shell welds and nozzle inner radius in NPPD Relief Request RI-04.
This submittal contains no new regulatory commitments.
If you have any questions concerning this matter, please contact David Van Der Kamp, Licensing Manager, at (402) 825-2904.
Sinc rel O4 r A. Limp s Vice President - Nuclear and Chief Nuclear Officer COOPER NUCLEAR STATION P.O. Box 98 / Brownville, NE 68321-0098 A 047 Telephone: (402) 825-3811 / Fax: (402) 825-5211 www.nppd.com
NLS2014029 Page 2 of 2
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Attachment:
10 CFR 50.55a Request Number RI-08, Revision 0 cc: Regional Administrator w/ attachment USNRC - Region IV Cooper Project Manager w/ attachment USNRC - NRR Project Directorate IV-1 Senior Resident Inspector w/ attachment USNRC - CNS NPG Distribution w/ attachment CNS Records w/ attachment
NLS2014029 Attachment Page 1 of 5 10 CFR 50.55a Request Number RI-08, Revision 0 Inspection of Reactor Vessel Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius Sections Cooper Nuclear Station Docket No. 50-298, DPR-46 Proposed Alternative in Accordance to 10 CFR 50.55a(a)(3)(i)
--Alternative Provides Acceptable Level of Quality and Safety--
COMPONENT IDENTIFICATION Code Class: I Examination Category: B-D (Inspection Program B)
Item Number: B3.90 and B3.100
==
Description:==
Inspection of Reactor Vessel Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius Sections Component Numbers:
Table RI-08-1 Specific Components that Relief is Requested Nozzle-to-Vessel Shell Welds that Relief is Requested ASME Item No. Component ID Item Description B3.90 NVE-BD-N2A 12" Recirculation Inlet B3.90 NVE-BD-N2B 12" Recirculation Inlet B3.90 NVE-BD-N2C 12" Recirculation Inlet B3.90 NVE-BD-N2D 12" Recirculation Inlet B3.90 NVE-BD-N2F 12" Recirculation Inlet B3.90 NVE-BD-N2G 12" Recirculation Inlet B3.90 NVE-BD-N2J 12" Recirculation Inlet Nozzle Inner Radius Sections that Relief is Requested ASME Item No. Component ID Item Description B3.100 NVIR-BD-N2A 12" Recirculation Inlet B3.100 NVIR-BD-N2B 12" Recirculation Inlet B3.100 NVIR-BD-N2C 12" Recirculation Inlet B3. 100 NVIR-BD-N2D 12" Recirculation Inlet B3.100 NVIR-BD-N2F 12" Recirculation Inlet B3.100 NVIR-BD-N2G 12" Recirculation Inlet B3.100 NVIR-BD-N2J 12" Recirculation Inlet
NLS2014029 Attachment Page 2 of 5
Applicable Code Edition and Addenda
The applicable Code Edition and Addenda for Cooper Nuclear Station (CNS) is American Society of Mechanical Engineers (ASME) Code Section XI, 2001 Edition, 2003 Addenda.
Additionally for ultrasonic examinations,Section XI, Appendix VIII, Performance Demonstration for Ultrasonic Examination Systems," is implemented as required and subject to the limitations and modifications specified in 10 CFR 50.55a(b)(2)(xv).
Applicable Code Requirement
ASME Section XI Code Class 1 nozzle-to-vessel shell weld and nozzle inner radius section examination requirements are provided in Subsection IWB, Table IWB-2500-1 "Examination Category B-D, Full Penetration Welded Nozzles in Vessels - Inspection Program B." Items B3.90 and B3.100 require a volumetric examination of all the Reactor Vessel nozzle-to-vessel welds and associated nozzle inner radius sections, respectively.
Reason for Request
The identified nozzles (see Table RI 1) are scheduled for examination prior to the end of the current inspection interval. The proposed alternative provides an acceptable level of quality and safety, and the reduction in scope could provide a dose savings of as much as 3.6 Radiation Equivalent Man over the remainder of the interval.
Proposed Alternative and Basis for Use Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested from performing the required examinations on 100% of the identified nozzle assemblies (see Table RI-08-2 below). As an alternative, CNS proposes to examine a minimum of 25% of the nozzle-to-vessel shell welds and nozzle inner radius sections in accordance with Code Case N-702 (Reference 3).
Table RI-08-2 Group/Component ID Total Number Nozzles No. to be Examined Recirculation Inlet 10 3 (N2)
Electric Power Research Institute (EPRI) Technical Report 1003557 (Reference 1), "BWRVIP-108: Boiling Water Reactor Vessel and Internals Project Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," provides the basis for Code Case N-702. The evaluation found that failure probabilities at the nozzle blend radius region and nozzle-to-vessel shell welds due to a Low Temperature Overpressure event are very low (i.e., < 1 x 10-6 for 40 years) with or without inservice inspection. The report concludes that inspection of 25% of each nozzle type is technically justified.
NLS2014029 Attachment Page 3 of 5 The Nuclear Regulatory Commission (NRC) issued a Safety Evaluation Report (SER) dated December 19, 2007 (Reference 2) approving the use of BWRVIP- 108 as a basis for using Code Case N-702. In the SER, Section 5.0, "Plant Specific Applicability" states that licensees who plan to request relief from the ASME Code,Section XI requirements for reactor pressure vessel (RPV) nozzle-to-vessel shell welds and nozzle inner radius sections may reference the BWRVIP-108 report as the technical basis for the use of ASME Code Case N-702 as an alternative.
Relative to BWRVIP- 108, CNS could not meet the NRC limitations for the recirculation N2 nozzles. However, EPRI Technical Report 1021005 (Reference 5), "BWRVIP-241 : Boiling Water Reactor Vessel and Internals Project Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-To-Vessel Shell Welds and Nozzle Blend Radii," provides the supplemental analyses for boiling water reactor (BWR) RPV recirculation inlet and outlet nozzle-to-vessel shell welds and nozzle inner radii to address limitations imposed by the NRC.
The NRC issued an SER dated April 19, 2013 (Reference 6), approving the use of BWRVIP-241 as a basis for using Code Case N-702. In the SER, Section 6.0, "Conclusions" states that licensees who plan to request relief from the ASME Code,Section XI requirements for RPV nozzle-to-vessel shell welds and nozzle inner radius sections may reference the BWRVIP-241 report as the technical basis for the use of ASME Code Case N-702 as an alternative. However, the licensees should demonstrate the plant specific applicability of the BWRVIP-241 report to their units in the relief request by addressing the conditions and limitations specified in Section 5.0 of the referenced SER.
The applicability of the BWRVIP-241 report to CNS is demonstrated by showing the criteria within Section 5 of the SER are met for the recirculation inlet nozzles.
Maximum RPV Heatup/Cooldown Rate (1) the maximum RPV heatup/cooldown rate is limited to less than 1157F per hour.
This criterion is met by adherence to CNS Technical Specifications Surveillance Requirement 3.4.9.1 which requires verification that the Reactor Coolant System heatup and cooldown rates are < 100'F when averaged over a one hour period.
For recirculation inlet N2 nozzles, Criteria (2) and (3) of the SER apply:
(2) (pr/t)/Ci-RPV < 1.15 where:
p = RPV normal operating pressure (p < 1020 psig per CNS Technical Specifications 3.4.10 for Reactor Steam Dome Pressure) r = RPV inner radius (r = 110.375 in.)
t = RPV wall thickness (t = 6.875 in.)
Ci-RPV = 19332 (based on the BWRVIP-108 recirculation inlet nozzle/RPV finite element method (FEM) model)
NLS2014029 Attachment Page 4 of 5 The calculation result is 0.847, which is less than 1. 15, therefore the CNS N2 nozzles meet Criteria 2.
(3) [p(ro2 +r, 2) / (ro2-r,2)] / Ci-NOZZLE < 1.47 where:
p = RPV normal operating pressure (p < 1020 psig) r0 = nozzle outer radius (ro = 10.219 in.)
ri = nozzle inner radius (ri= 6.188 in.)
Ci-NOZZLE = 1637 based on the BWRVIP-108 recirculation inlet nozzle/RPV FEM model)
The calculation result is 1.344, which is less than 1.47, therefore the CNS N2 nozzles meet Criteria 3.
Based upon the above information, all RPV nozzle-to-vessel shell welds and nozzle inner radius sections for which relief is requested meet the criteria.
Therefore, use of Code Case N-702 provides an acceptable level of quality and safety pursuant to 10 CFR 50.55a(a)(3)(i) for the RPV N2 nozzle-to-vessel shell welds and nozzle inner radius sections as described above.
Duration of Proposed Alternative The proposed alternative is for the fourth ten-year interval of the Inservice Inspection Program for CNS that started on March 1, 2006 and ends on February 29, 2016.
Precedent This request is similar to the submittal the NRC Staff has approved in a Safety Evaluation for other nozzle-to-vessel shell welds and nozzle inner radius in Relief Request RI-04 for CNS (Reference 4).
References
- 1. EPRI Technical Report 1003557, "BWRVIP-108: Boiling Water Reactor Vessel and Internals Project Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," October 2002.
- 2. NRC Safety Evaluation of Proprietary EPRI Report, "BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius (BWRVIP- 108)," December 19, 2007.
- 3. ASME Boiler and Pressure Vessel Code,Section XI Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell WeldsSection XI, Division I ," dated February 20, 2004.
NLS2014029 Attachment Page 5 of 5
- 4. Cooper Nuclear Station - Request for Relief No. RI-04 for the Fourth 10-Year Inservice Inspection Interval Regarding Inspection of Reactor Vessel Nozzle-to-Vessel Shell Welds (TAC No. ME3319), dated October 8, 2010.
- 5. EPRI Technical Report 1021005, "BWRVIP-241: BWR Vessel and Internals Project Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," October 2010.
- 6. Final Safety Evaluations of the Boiling Water Reactor Vessel Internals Project (BWRVIP)-
241 Report, "Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii" (TAC NO. ME6328), April 19, 2013.